1. Field of the Invention
This invention relates to a reactor core monitoring system and method, and especially to a monitoring system which monitors the critical power ratio of each of the fuel segments in an atomic reactor.
2. Background of the Invention
In order to ensure the health of the fuel during operation of an atomic reactor, burn-out of the fuel rods must be prevented. This is done by preventing a shift from a nucleate boiling condition to a boiling transition condition caused by heating, from the fuel rods, of the cooling water used to cool the fuel assembly. The critical power ratio (CPR) is employed as an index of boiling transition. EQU CPR=critical power (CP)/channel power . . . (1)
where the critical power CP is defined as the power of the fuel assembly that generates boiling transition.
The critical quality Xc is used as the index of critical power at this point. The critical quality Xc is calculated using the boiling transition correlation expression. In the case of a boiling water reactor, the following GEXL expression given for example in "The thermal-hydraulics of a boiling water nuclear reactor", by R. T. Lahey, Jr. and F. J. Moody, published by American Nuclear Society, second printing, 1979, is usually employed. EQU Xc=f(LB, Da, G, L, Pr, R) . . . (2)
where
LB=boiling length PA1 Da=thermal equivalent diameter PA1 G=coolant mass flux PA1 L=heat length PA1 Pr=pressure PA1 R=a factor characterizing the local power distribution within the fuel assembly.
The R-factor is included in this correlation expression (1) for the critical quality Xc. This R factor is defined as a function of the local power distribution p, and is defined for each individual fuel rod.
Here, the local output power is the power density of each fuel rod when the average power of all the fuel rods of the fuel assembly is normalized to 1.0. That is, the R factor of the fuel assembly is given as a function of the most limiting fuel rod power density or the power density of the peripheral fuel rods. EQU R=R(p) . . . (3)
In monitoring the power distribution or predicting the power distribution of the core, the power of the assembly is usually calculated by the following method. Specifically, in general the fuel assembly constituting the core is divided into a large number of fuel segments in the axial direction, and the approximation is made that the fuel composition is homogeneous within these fuel segments The few-group neutron diffusion equation or modified-one-group neutron diffusion equation is then solved, to calculate the average neutron flux taking a fuel segment as a unit and the thermal power density for each of the fuel segments. This is a homogeneous neutron diffusion calculation taking into account the movement of neutrons between the respective fuel segments, and taking into account the effects of fuel type, void fraction, and exposure, etc., of neighboring fuel segments. However, within each segment, the neutron flux and thermal power density are calculated on the assumption that these are homogeneous. Consequently, the local power distribution or R factor within a fuel segment, which depends on the heterogeneity of composition within the fuel segment, cannot be found.
The distribution of neutron flux and thermal power density for each of the fuel segments which are thus obtained express the global power distribution of the core as a whole. They are therefore termed the global power distributions.
In general, the local power distribution within a fuel segment depends on a characteristic index of the fuel segment determined by the fuel type, for example the geometrical shape of the fuel segment itself, the density and degree of enrichment of the charged fuel, the distribution, etc., of burnable poisons such as gadolinia, the void fraction of the fuel segment itself, and exposure. It also depends on the fuel type, void fraction, and exposure of fuel segments near to the fuel segment under consideration, and on the pattern of insertion of neighboring control rods. This is because during operation neutrons move between fuel segments in the core.
Consequently, although a homogeneous neutron diffusion calculation based on the approximation that the composition of the material within the fuel segments is homogeneous does take into account the effect of neighboring fuel segments, it cannot be used to find the local power distribution or R factor within a fuel segment.
Consideration has therefore been given to calculating the local power distribution within a fuel segment by a heterogeneous neutron diffusion calculation taking into account the movement of neutrons between fuel segments in the core. However, this is not practical as it requires a lot of computational cost and time.
The local power distribution and R factor within a fuel segment were therefore found by a heterogeneous neutron diffusion calculation in an infinite lattice consisting only of the fuel segment in question. An "infinite lattice" is a single fuel segment system to which a mirror symmetry boundary condition is applied at the boundaries in all directions of the fuel segment in question. This heterogeneous neutron diffusion calculation in an infinite lattice was usually carried out at the design stage of the fuel assembly.
In this heterogeneous neutron diffusion calculation in an infinite lattice, the local power distribution taking into account heterogeneity within the fuel segment in obtained, but the effect of neighboring fuel segments is not considered.
However, with fuel improvements made in recent years with the object of raising core performance, it has become common for different types of fuel to be arranged adjacent to each other in the core. This has increased the effect on the local power distribution. This has led to the problem in monitoring the core that since, in finding the R factor by a heterogeneous neutron diffusion calculation in an infinite lattice, the effects of fuel type, void fraction, and exposure of neighboring fuel segments on the above-mentioned local power distribution and the effect of neighboring control rods are neglected, monitoring of the reactor is more difficult due to a lowering of the accuracy of determining the local power distribution. The R factor must be taken into consideration.